Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

Federal RegisterMar 2, 1994

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from February 5, 1994, through February 17, 1994.

The last biweekly notice was published on February 16, 1994 (59 FR

7685).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue,

Bethesda, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies

of written comments received may be examined at the NRC Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555.

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By April 1, 1994, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

public document room for the particular facility involved. If a request

for a hearing or petition for leave to intervene is filed by the above

date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

and Licensing Board Panel, will rule on the request and/or petition;

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the

above date. Where petitions are filed during the last 10 days of the

notice period, it is requested that the petitioner promptly so inform

the Commission by a toll-free telephone call to Western Union at 1-

(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and

to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC 20555, and at the local public document

room for the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit

Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: January 13, 1994

Description of amendment requests: Request for NRC consent to the

indirect transfer of control of El Paso Electric Company's interest in

Operating License Nos. NPF-41, NPF-51, NPF-74 and to amend Operating

License Nos. NPF-51 and NPF-74 to delete provisions for El Paso

Electric Company's sale-leaseback arrangements.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1 -- Involve a significant increase in the probability

or consequences of an accident previously evaluated.

This amendment request does not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the proposed change is administrative in nature.

The proposed change deletes Sections 2.B.(7)(a) and (b) of License

No. NPF-51, and Sections 2.B.(6)(a) and (b) of License No. No. NPF-

74. These section describe the structure of the financing of El

Paso's interest in Palo Verde, specifically authorizing sale and

leaseback transactions. The proposed change does not affect the

assumptions used in the accident analyses, nor does the proposed

change result in changes to the physical configuration of the

facility, design parameters, technical specifications, or operation

and maintenance of the facility. Therefore, this amendment request

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Standard 2 -- Create the possibility of a new or different kind

of accident from any accident previously analyzed.

This amendment request does not create the possibility of a new

or different kind of accident from any accident previously analyzed

because the proposed change is administrative in nature. The

proposed change deletes Sections 2.B.(7)(a) and (b) of License No.

NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These

sections describe the structure of the financing of El Paso's

interest in Palo Verde, specifically authorizing sale and leaseback

transactions. The proposed change does not involve modifications to

any of the existing equipment nor does the change affect the

operation and maintenance of the facility. Therefore, this amendment

request does not create the possibility of a new or different kind

of accident not previously analyzed.

Standard 3 -- Involve a significant reduction in a margin of

safety.

This amendment request does not involve a significant reduction

in a margin of safety because it is administrative in nature. The

proposed change deletes Sections 2.B.(7)(a) and (b) of License No.

NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These

sections describe the structure of the financing of El Paso's

interest in Palo Verde, specifically authorizing sale and leaseback

transactions. The proposed change does not involve changes to any

existing plant equipment or accident analyses that provide for or

establish margins of safety. There are no changes to the operation

or maintenance of the facility and the existing margins of safety

are not changed by the proposed change. Therefore, this amendment

request does not involve a signigicant reduction in a margin of

safety.

The NRC staff has reviewed the licensees' analysis and, based on

that review, it appears that the proposed license amendment reflects

only a change in the structure of the financing of El Paso's interest

in Palo Verde and the three standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to determine that the amendment

requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004 Attorney for licensees:

Nancy C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public

Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona

85072-3999

NRC Project Director: Theodore R. Quay

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of amendment request: February 4, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification 3/4.6.4, Containment Systems Combustible

Gas Control, by eliminating the 12-hour channel check surveillance

requirement for the containment hydrogen monitoring system in

conformance with the new Standard Technical Specifications, NUREG-1431.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Final Safety Analysis Report [FSAR] section 6.2.5.2.3 states

that the Hydrogen Analyzer is only required to be functioning

(continuously indicating and recording hydrogen concentration)

within 30 minutes of safety injection initiation. The performance of

an analog operational test every 31 days and a channel calibration

test every 92 days verifies this operability. Based on this, the

monitors will be fully capable of performing their intended design

function following a safety injection initiation. Therefore, the

elimination of the 12-hour channel check would not increase the

probability or consequences of an accident previously evaluated.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The Hydrogen Monitors perform an ``indication'' function only,

[sic] to help ensure that hydrogen concentrations within containment

are maintained below flammable limits during a post-LOCA [loss-of-

coolant accident] condition. The proposed changes do not involve any

modifications or additions to plant equipment and the design and

operation of the plant will not be affected. Therefore, the

elimination of the 12-hour channel check does not affect any

parameters which relate to the margin of safety as defined in the

Technical Specifications or the FSAR. Therefore, the proposed

changes do not involve a significant reduction in a margin of

safety.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed elimination of the 12-hour channel check does not

affect any parameters which relate to the margin of safety as

defined in the Technical Specifications or the FSAR. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: S. Singh Bajwa

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: March 26, 1993

Description of amendment request: The proposed amendment would

modify trip level settings for the Isolation Condenser and High

Pressure Core Injection (HPCI) System Steam lines to more conservative

values. In addition, the proposed amendment would revise the Emergency

Core Cooling System Low-Low Water Level initiation trip level setting

tolerance.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

HPCI Steamline High Flow Isolation Trip Level Setting

The purpose of the HPCI leak detection systems are to detect

breaks in the system piping. Normal steam flows within the system

can fluctuate in excess of 250% rated flow and exceed 500% rated

steam flow after experiencing a break. During the original licensing

of the plant, it was analytically determined by GE that three times

maximum steam flow (300%) is the optimum setpoint for the isolation

of HPCI. A 300% steam flow setpoint ensures that spurious trips are

avoided and that breaks in the piping are identified. Because the

HPCI High Steamline Flow Isolation setpoint is not assumed as an

accident precursor, the probability of any previously evaluated

accident is not increased by the changed setpoint.

The proposed changes to the setpoint allow a more accurate and

conservative value for 300% steam flow. The proposed change in

conjunction with a more conservative field setting ensures HPCI

isolation occurring between the range of 300% and 500% steam flow,

thus ensuring HPCI isolation in the event of a pipe break. Because

the HPCI high steamline flow setpoint will be maintained above

normally found operational values (272% steam flow) and below

expected conditions with a pipe break (500% steam flow), the

consequences of any previously evaluated accident are not increased

with the proposed setpoint change.

solation Condenser Steamline High Flow Isolation Trip Level

Setting

The purpose of the Isolation Condenser leak detection

instrumentation is to detect breaks in the system piping. Normal

steam flows within the system can fluctuate in excess of 250% rated

flow and exceed 500% rated steam flow after experiencing a break.

During the original licensing of the plant, it was analytically

determined by GE that three times rated steam flow (300%) is the

optimum setpoint for the isolation of the Isolation Condenser. A

300% steam flow Isolation setpoint ensures that spurious trips are

avoided and that breaks in the piping are identified. Because the

Isolation Condenser High Steamline Flow setpoint is not assumed as

an accident precursor, the probability of any previously evaluated

accident is not increased by the changed setpoint. The proposed

changes to the setpoint provide a more accurate and conservative

field setting for 300% steam flow.

The proposed changes in conjunction with a more conservative

field setting results in Isolation Condenser isolation occurring

between the range of 300% and 500% steam flow, thus ensuring

Isolation Condenser isolation in the event of a pipe break. Because

the Isolation Condenser High Steamline Flow Isolation setpoint will

be maintained above normally found operational values (272% steam

flow) and below expected conditions with a pipe break (500% steam

flow), the consequences of any previously evaluated accident are not

increased with the proposed setpoint change.

Reactor Low-Low Water Level Trip Level Setting Tolerance

The Low-Low Reactor Water Level trip setting is designed to

initiate ECCS when reactor water level is less than or equal to 444

inches above vessel zero. Top of active fuel (TAF) is defined as 360

inches above vessel zero. -59 inches is 84 inches above TAF. The

present trip setting tolerance (84 inches, + 4, - 0, above TAF) only

allows a deviation of 4 inches in the conservative direction. The

proposed change (greater than or equal to 84 inches) does not impose

a restriction on the limit toward the conservative direction.

Because a level switch trip level setting by itself is not assumed

as an accident precursor, the probability of any previously

evaluated accident is not increased by the changed setpoint.

The proposed change eliminates a restriction on the trip level

setting for Low-Low Reactor Water Level. Dresden proposes modifying

the acceptance limit of the Low-Low trip setting such that the

instrument field setting will not deviate below 84 inches.

Therefore, the actuation of appropriate ECCS are unchanged and the

consequences of any previously evaluated accident are not increased

with the proposed setpoint change.

Create the possibility of a new or different kind of accident

from any accident previously evaluated because:

HPCI Steamline High Flow Isolation Trip Level Setting

The purpose of the HPCI Steamline High Flow Isolation trip level

setting is to detect breaks in system piping and initiate isolation

of the system if breaks are discovered. Normal steam flows within

the system can fluctuate as high as 250% rated flow and exceed 500%

rated steam flow after experiencing a break. 300% steam flow has

been used as the setpoint to ensure that spurious trips are avoided

and that breaks in the piping are identified. The changes to the

HPCI High Steamline Flow setpoint ensure that isolation occurs at

300% rated steam flow (below 500% rated steam flow). The current

setpoint will also isolate below 500% rated steam flow. Because the

new setpoint continues to allow normal operational flexibility and

ensures isolation in the event of a pipe break, the proposed changes

do not create the possibility of a new or different kind of accident

than previously evaluated.

Isolation Condenser Steamline High Flow Isolation Trip Level

Setting

The purpose of the Isolation Condenser Steamline High Flow

Isolation trip level setting is to detect breaks in system piping

and initiate isolation of the system if breaks are discovered.

Normal steam flows within the system can fluctuate in excess of 250%

rated flow and exceed 500% rated steam flow after experiencing a

break. 300% steam flow has been used as the setpoint to ensure that

spurious trips are avoided and ensures that isolation occurs at 300%

rated steam flow (below 500% rated steam flow). The current setpoint

will also isolate below 500% rated steam flow. Because the new

setpoint continues to allow normal operational flexibility and

ensures isolation in the event of a pipe break, the proposed changes

do not create the possibility of a new or different kind of accident

than previously evaluated.

Reactor Low-Low Level Trip Level Setting Tolerance

The Reactor Low-Low Water Level trip setting is designed to

initiate the appropriate ECCS when Reactor Water Level is

decreasing. The proposed change to the setpoint only eliminates the

overly burdensome restriction within the setpoint tolerances. The

absolute low limit of 84 inches is unchanged, thus maintaining all

assumptions related to 84 inches (-59 inches indicated level) within

Dresden's Safety Analysis. The removal of the upper tolerance will

not increase the probability of inadvertent ECCS initiation since

the actual field setting will be at a reactor vessel level which has

not been reached in 40 + years of operation at Dresden Units 2 and

3. Therefore, the proposed changes do not create the possibility of

a new or different kind of accident than previously evaluated.

Involve a significant reduction in the margin of safety because:

High Pressure Coolant Injection Setpoint

The HPCI high steamline flow setpoint ensures that isolation

occurs at 300% maximum steam flow (below 500% rated steam flow). The

current Technical Specification setpoint will also allow isolation

below 500% rated steam flow but at a value greater than 300%. Thus,

the proposed setpoint isolates at a lower steam flow rate than the

current limit. Therefore, because isolation of HPCI would occur at a

lower steam flow rate during a pipe break, the proposed changes do

not involve a significant reduction in the margin of safety.

Isolation Condenser Steamline High Flow Isolation Trip Level

Setting

The Isolation Condenser High Steamline Flow Isolation Trip Level

setting ensures that isolation occurs at 300% rated steam flow

(below 500% rated steam flow). The current setpoint will also

isolate below 500% rated steam flow but at a value greater than

300%. Thus, the proposed setpoint isolates at a lower steam flow

rate than the current limit. Therefore, because isolation of the

Isolation Condenser would occur at a lower steam flow rate during a

pipe break, the proposed changes do not involve a significant

reduction in the margin of safety.

Reactor Low-Low Level Trip Level Setting Tolerance

The Reactor Low-Low Water Level trip setting tolerance ensures

the proper initiation of appropriate ECCS in the event of a loss of

inventory to the vessel. The proposed change to the setpoint only

eliminates the restriction within the setpoint tolerances. The

absolute low limit of 84 inches is unchanged, thus maintaining all

assumptions related to 84 inches (minus 59 inches indicated) within

Dresden's Safety Analysis. Therefore, the proposed changes do not

involve a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Morris Public Library, 604

Liberty Street, Morris, Illinois 60450

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: James E. Dyer

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1 and 2, Rock Island County,

Illinois

Date of amendment request: December 20, 1993

Description of amendment request: The proposed amendment would

revise a minimum critical power ratio (MCPR) safety limit from 1.06 to

1.07 based on General Electric Standard Application for Reactor Fuel II

(GESTAR II) NEDE-24011-P-A-10 for GE10 fuel design. The NRC staff has

previously reviewed and approved the GE10 fuel design.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

The change is based on GE's generic rule licensing document

GESTAR II (NEDE-24011-P-A-10) which has conservatively addressed the

use of GE10 fuel in D-lattice cores with NRC approved methods and

therefore does not adversely affect the consequences of previously

evaluated accidents. The Safety Limit MCPR change does not affect

the probability of analyzed accidents because it does not adversely

impact any equipment important to safety. Increasing the Safety

Limit MCPR from 1.06 to 1.07 upon implementation of GE10 fuel for

Cycle 14 operation of Quad Cities Units 1 and 2 therefore does not

involve a significant increase in the probability or consequences of

any accident previously evaluated in the FSAR.

The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated

because:

The Safety Limit MCPR change results from the use of NRC

approved methods in GESTAR II NEDE-24011-P-A-10 for application to

GE10 fuel for Cycle 14 for Quad Cities Units 1 and 2. The Safety

Limit MCPR change does not result in any new interaction with

equipment related to the safe shutdown of the plant. The change does

not adversely impact equipment important to safety and, therefore

does not create the possibility of a new or different kind of

accident scenario. Therefore, the Safety Limit MCPR change from 1.06

to 1.07 in no way creates the possibility of a new or different kind

of accident scenario from any accident previously evaluated.

The proposed change does not involve a significant reduction in

a margin of safety because:

Since the GE10 design in a D-lattice core has a geometry between

C-lattice and D-lattice designs and the C-lattice design has a

higher, more restrictive safety limit MCPR that the D-lattice

design, the use of C-lattice safety limit MCPR for the GE10 design

is a conservative approach. The GE10 fuel design has been

generically analyzed with approved methods per GESTAR II NEDE-24011-

P-A-10 and the use of the 1.07 Safety Limit MCPR value has been

previously approved as conservative for application to GE10 fuel in

D-lattice plants such as Quad Cities. Therefore, the proposed change

to increase the Safety Limit MCPR from 1.06 to 1.07 maintains the

margin to safety relative to the current level.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: James E. Dyer

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: December 10, 1993

Description of amendment request: The proposed amendment request

would revise the Technical Specifications to amend (1) Section 5.3.A

(Reactor Core) to allow the use of VANTAGE + fuel with ZIRLO cladding

and fuel with filler rods to allow fuel reconstitution, and (2) the

Basis to Section 2.1 (Safety Limit: Reactor Core) to allow the use of

departure from nucleate boiling (DNB) Correlations applicable to

VANTAGE + fuel.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Consistent with the requirements of 10 CFR 50.92, the enclosed

application involves no significant hazards based on the following

information:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

Neither the probability nor the consequences of an accident

previously analyzed is increased due to the proposed changes. As

discussed in [Letter from Thadani to Tritch, ``Acceptance

for Referencing of Topical Report WCAP-12610, VANTAGE + Fuel

Assembly Reference Core Report'' (TAC No. 77258) July 1, 1991] the

fuel containing ZIRLO clad will meet all the same material and

mechanical design criteria as the Zircaloy clad fuel. The use of

approved Westinghouse Methodology for fuel assembly reconstitution

as documented in [Letter from Thadani to Tritch, ``Acceptance for

Referencing of Topical Report WCAP-13060-P, Westinghouse Fuel

Assembly Reconstitution Evaluation Methodology'' (TAC No. M821391),

March 30, 1993] will ensure that all criteria are met. The change to

the basis of Section 2.1 more accurately describes DNB methodology

and application.

2. Does the proposed license amendment create the possibility of

a new or different kind of accident from any previously evaluated?

Response:

The changes will not create the possibility of a new or

different kind of accident. The proposed changes involve approved

methodology which have been shown to meet design and safety

criteria. In addition, approved procedures will be used to implement

the changes.

Response:

3. Does the proposed amendment involve a significant reduction

in the margin of safety?

The proposed amendment does not involve a significant reduction

in the margin of safety. The changes involve the use of approved

methodology which meet design and safety criteria. The change to the

Section 2.1 basis is descriptive and will more accurately describe

the DNB methodology used in conjunction with the use of VANTAGE +

fuel.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: Robert A. Capra

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: September 28, 1993

Description of amendment request: The requested amendments would

delete the portion of the 18-month surveillance requirement on the

autoclosure interlock (ACI) contained in TS 4.5.2.d associated with

verifying that the decay heat removal system suction isolation valves

automatically close on a reactor coolant system pressure signal. The

terms decay heat removal (ND) and residual heat removal (RHR) are used

interchangeably here. Also, an obsolete footnote to TS 4.5.2.e relating

to the completion of the first Unit 1 refueling outage is proposed to

be deleted.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The requested amendments reference Westinghouse topical report

WCAP-11736-A, ``Residual Heat Removal System Autoclosure Interlock

Removal Report for the Westinghouse Owners Group'', for the general

justification and safety analysis for removing the ACI feature from

the Catawba ND suction isolation valves. This WCAP, which

specifically covers the Catawba Nuclear station, has been deemed an

acceptable reference by the NRC for use in making plant-specific

licensing submittals. Additional Catawba-specific information/

improvements and analyses, as required by the WCAP and associated

NRC safety evaluation, have been either completed or committed to,

thereby ensuring that the WCAP/SE conclusion that removal of the RHR

ACI produces a net safety benefit remains valid.

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The deletion of the RHR ACI was analyzed in the WCAP for

Callaway Nuclear Station in terms of (1) the frequency of an

interfacing LOCA, (2) the availability of the RHR system, and (3)

the effect on overpressure transients. Callaway is the WCAP's

reference plant for Catawba Units 1 and 2, and a Catawba-specific

Probabilistic Risk Assessment (PRA) review of the WCAP determined

that removal of the ND ACI at Catawba will not invalidate the basic

conclusions of the WCAP. Consequently, the following information

from the Callaway analysis is considered applicable to Catawba Units

1 and 2.

With the removal of the ACI and addition of a control room

alarm, the probabilistic risk analysis predicts a decrease in the

frequency of interfacing LOCAs from 1.52E-06/year to 1.16E-06/year,

a decrease of approximately 24%.

The availability of the RHR system was analyzed in three phases:

initiation, short term cooling, and long term cooling. The

probabilistic analysis indicated that deletion of the RHR ACI has no

impact on the failure probability for RHR initiation. During short

term cooling (72 hours after initiation), RHR ACI deletion decreased

the RHR failure probability by 12%, from 1.64E-02 to 1.44E-02. The

long term cooling RHR failure probability was calculated to decrease

by 70%, from 3.91E-02 to 1.17E-02.

Appendix D of the WCAP presents the analysis used to determine

the effect of removal of the ACI on overpressurization transients.

The analysis categorizes the types of initiating events, determines

their frequency of occurrence, and then identifies the consequences

of these occurrences both with and without the ACI feature. The

result is a list of overpressure consequence categories with

associated failure probabilities (reference the WCAP's Appendix D,

Tables D-14, -15, and -16). For the charging/safety injection event,

consequence frequencies increased on the order of 1.0E-12/shutdown

year. This is an insignificant increase, as the overall consequence

frequency of the charging/safety injection event is 1.25E-01.

Likewise, for the letdown isolation with RHR system operable case,

one frequency category was increased on the order of 1.0E-15. Again,

this is insignificant when compared with the total frequency of

these events of 1.25E-01. For the letdown isolation with RHR system

isolated event, the overall consequence frequency was reduced from

4.45E-01 to 2.22E-01. This occurs because many spurious closures of

the RHR isolation valves cause the isolation of letdown. Removing

the RHR ACI reduces the frequency of this event by approximately

50%. It is concluded that the removal of the RHR ACI circuitry has

an insignificant impact on the frequency of overpressurization

events at Callaway (and thus Catawba) Nuclear Station.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. The effect of an overpressure transient at cold shutdown

conditions will not be altered by removal of the ND ACI function.

With or without the ACI function, the ND system could be subject to

overpressrue for which the ND relief valves must be relied upon to

limit pressure to within ND design parameters. While it is true that

the ACI initiates an automatic closure of the ND suction/isolation

valves on high NC system pressure, overpressure protection of the ND

system is provided by the ND system relief valves and not by the

suction/isolation valves that isolate the ND system from the NC

system. (Refer to NUREG-0954, ``Safety Evaluation Report related to

the operation of Catawba Nuclear Station, Units 1 and 2,'' Section

5.4.4.3.)

The purpose of the ACI feature is to ensure that there is a

double barrier between the ND system and the NC system when the

plant is at normal operating conditions (i.e., heated and

pressurized) and not in the ND cooling mode. Thus, the ACI feature

serves to preclude conditions that could lead to a LOCA outside of

containment due to operator error. The safety function of the ACI is

not to isolate the ND system from the NC system when the ND system

is operating in the decay heat removal mode.

There are several methods to ensure that there is a double

barrier between the ND system and the NC system when the plant is at

normal operating conditions. First, plant operating procedures

instruct the operators to isolate the ND system during plant heatup.

Second, the alarm that will be installed as part of this change will

annunciate in the control room given an open or intermediate valve

position signal in conjunction with a high NC pressure signal. This

alarm will alert operators that any of the four suction/isolation

valves is (are) not fully closed and that double isolation has not

been achieved. In conjunction with this alarm, operators will be

trained using an annunciator response procedure to ensure that they

act to restore double isolation or return to a safe shutdown

condition. Third, the Open Permissive Interlock (OPI), which is not

being removed, will prevent the opening of the valves whenever NC

system pressure is greater than 385.5 psig.

Since relief valves prevent overpressurization of the ND system

during shutdown conditions and since several methods are in place to

ensure that the ND system is isolated from the NC system during

normal plant conditions, removal of the ACI will not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. The ND ACI function is not a

consideration in a margin of safety in the basis for any technical

specification. Since the probabilistic analysis of the WCAP for

Callaway (which is applicable to Catawba as discussed above)

indicates that the availability of the RHR system is increased with

the removal of the ACI, overall safety will be increased.

In addition, similar amendments for other Westinghouse plants in

the past have been determined to not involve significant hazards

considerations.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Loren R. Plisco, Acting

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: January 27, 1994

Description of amendment request: The requested amendments delete

the verification that each upper and lower Containment Purge System

(VP) supply and exhaust valve actuates to its isolation position on a

High Relative Humidity (70%) isolation test signal and will

allow elimination of the humidity control function of the VP System

humidistats.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

CRITERION 1

This TS [Technical Specification] amendment will not increase

the probability or consequences of an accident which has been

previously evaluated. No physical changes will be made to the plant

that would impact fuel handling inside containment, therefore, there

is no increase in the probability of an accident. Control wiring

changes that remove the humidistats from the [Containment Purge]

System control circuits will be the only physical change.

The heaters will be maintained providing additional margin over

analyzed conditions. For the reasons stated above, there will be no

increase in the consequences of an accident previously evaluated.

CRITERION 2

This proposed TS amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated. This proposed TS amendment will not cause any physical

changes to the plant that will impact the handling of fuel inside

containment or changes to fuel handling procedures. Because the

plant will operate the same way it does now, this proposed amendment

does not create the possibility of any new or different accident

from any previously evaluated.

CRITERION 3

This proposed TS change will not cause a significant reduction

in the margin of safety. The test method use[d] to evaluate the

carbon after TS changes 90 ([Unit] 1) and 84

([Unit] 2) does not consider heater availability. However the

heaters will be tested and maintained per Technical specification

4.9.4.2.d.2. Therefore, the relative humidity of the air entering

the carbon adsorber is never expected to reach 95% [relative

humidity].

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Loren R. Plisco, Acting

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit No. 1 (ANO-1), Pope County, Arkansas

Date of amendment request: January 13, 1994

Description of amendment request: This amendment revises the

specifications governing the reactor protection (RPS). It modifies the

use of the RPS channel bypass as specified by Technical Specification

(TS) 3.5.1.3 and revises a note with Table 3.5.1-1, to refer to a more

appropriate action.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The RPS and EFIC [emergency feedwater initiation and control]

system provide accident mitigation features and are not considered

to be accident initiators. The accident mitigation features of the

plant are not affected by the proposed amendment. In any

configuration allowed by the revised specifications, the trip logic

instituted on the RPS is at least equivalent to the trip logic

instituted by placing a channel in channel or maintenance bypass.

The RPS remains single-failure proof with one channel in channel

bypass, manually tripped, or with an inoperable function unbypassed

in the untripped state. Therefore, upon receipt of an initiating

signal, a single failure will not prevent the proper actuation of

RPS. Should a channel of RPS contain an inoperable function

unbypassed in the untripped condition which does not affect an EFIC

channel, any channel of EFIC may be placed in maintenance bypass.

RPS and EFIC remain single-failure proof in this configuration.

Administrative controls are established to ensure that all

inoperable RPS functions are evaluated for continued operation in

the untripped state. Upon detection of a failed function in any

channel of RPS, the administratively controlled condition reporting

process evaluates the failure and its effect on other systems for

continued operability. The operator is informed of the continuing

status of inoperable functions through the use of Station Log

entries and Plant Status board entries. In addition, during

operation with an inoperable function in the untripped, unbypassed

condition, the remaining RPS channel key-lock channel bypass

switches will be ``Hold Carded'' (tagged) to prevent their operation

without prior management approval consistent with the requirements

of TS Table 3.5.1-1. Plant management maintains the responsibility

to approve continued operation with inoperable functions unbypassed

in the untripped state to ensure that the plant is operated in the

safest configuration with regard to the extent of the failure, and

the plant operating conditions. Prior to placing any channel of RPS

or EFIC in bypass, the operator checks the status of redundant

systems for operability and TS compliance and takes the proper

action as required by existing plant conditions, plant operating

procedures and TS.

The clarification to TS 3.5.1.3 which directs the operator to

the appropriate actions if multiple channels become inoperable, or

in the event of an inoperable channel or inoperable function

occurring concurrent with one channel in bypass is considered to be

administrative in nature. The change to Note 6 of Table 3.5.1-1

results in the correction of misleading information and directs the

Operator to place the plant in a safe mode depending on the system

which is affected by a failure, and is also considered to be

administrative in nature. The Bases changes add additional

information to clarify the specifications.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The probability or consequences of equipment important to safety

malfunctioning will not be increased. In any configuration allowed

by the revised specifications, the trip logic instituted on the RPS

is at least equivalent to the trip logic instituted by placing a

channel in channel bypass. The RPS remains single-failure proof with

one channel in channel bypass, manually tripped, or with an

inoperable function unbypassed in the untripped state. Therefore,

upon receipt of an initiating signal, a single failure will not

prevent the proper actuation of RPS. Should a channel of RPS contain

an inoperable function unbypassed in the untripped condition which

does not affect an EFIC channel, any channel of EFIC may be placed

in maintenance bypass. RPS and EFIC remain single-failure proof in

this configuration.

The clarification to TS 3.5.1.3 which directs the operator to

the appropriate actions if multiple channels become inoperable, or

in the event of an inoperable channel or inoperable function

occurring concurrent with one channel in bypass is considered to be

administrative in nature. The change to Note 6 of Table 3.5.1-1 is

also considered to be administrative in nature, in that misleading

information in the specification has been corrected to an

appropriate requirement. The Bases changes add additional

information to clarify the specifications.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

The RPS and EFIC system have the same capabilities to mitigate

and/or prevent accidents as they had prior to this proposed change.

Allowing flexibility in the response to a function failure in one

channel of RPS allows placing the plant in the safest operating

condition for the existing plant conditions considering the extent

of the function failure. Operation of an RPS channel with an

inoperable function unbypassed in the untripped state results in

placing the inoperable function in a 2-out-of-3 trip logic

(equivalent to channel bypass) while the remainder of the RPS

functions remain in the normal 2-out-of-4 trip logic. The ANO-1 RPS

has been reviewed as a 3 channel system with one channel in bypass.

Implementing this change results in additional conservatism with

respect to any postulated single-failures.

Administrative controls are established to ensure that all

inoperable RPS functions are evaluated for continued operation in

the untripped state. Upon detection of a failed function in any

channel of RPS, the administratively controlled condition reporting

process evaluates the failure and its effect on other systems for

continued operability. The operator is informed of the continuing

status of inoperable functions through the use of Station Log

entries and Plant Status board entries. In addition, during

operation with an inoperable function in the untripped, bypassed